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R$7UELING,SEERATIONS .

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-3/4.9.6__ REFUELING PLATF0PM

LIMITING CONDITIONS FOR OPERATIONt

3.9.6 The refueling platform shall be OPE.RABLE and used for handling' fuelassemblies or control rods within the reactor vessel.

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AffLICARILIII: During handling of fuel assemblies or control rods within thereactor pressure vessel.

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Vith the requirements for refueling platform OPERABILITY not satisfied,suspend use of any inoperable refueling platform equipment from operations j

involving the handling of control rods and fuel assemblies within the reactorpressure vessel after placing the load in a safe condition.

SURVEIptANCEREQUIREMENTS __ ____ ,

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4.9.6 Each refueling platform crane or hoist used for handling of control rodsor fuel assemblies within the reactor pressure vessel shall be demonstrated <

OPERABLE vithin 7 days before.the start of such operations vivh that crane or i

hoist by::

Demonstrating operation-of the overload cutoff on the main hoist whena.the load exceeds 1200 50 pounds,

b. Demonstrating operation of the overload cutoff on the frame mounted and ,

monorail mounted auxiliary holsts when the load exceeds 1000 50pounds. .

5Demonstrating operation of'the main and auxiliary hoist uptravel stopsc.when the grapple is lover.than or equal to 7 feet 3 3/4 inches below the | ;

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platform tracks..

td. ; Demonstrating operation of the downtravel mechanical _ cutoff on the main

hoist when grapple hook down travel reaches 4-inches below fuel assembly-handle. ;

Demonstrating operation of the slack cable cutoff on the main hoist when '

e.the load is less than 50 10 pounds. .

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f. Demonstrating operation of the loaded interlock on the main hoist when '

the load exceeds 485 50 pounds,

g.. Demonstrating operation of the redundant. loaded interlock on the mainhoist when- the load exceeds 550 50 pounds. .

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9103300177 910821 |

PDR ADOCK 05000410'

P PDR .

NINE MILE POINT - UNIT 2 3/4 9-8 -

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ATTACSEENT B-*

NIAGARA MORAW POVER ColltP0 RATION

LICENSE NO. NPF-69

DOW ET No. 50-410

Sgpportiun InforRLi9D_and No .S. igd 1.UXani_JAlfJJA_fenRidgIA1.19D Analvala3

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1.0 Qgseriotion of Changas~

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The proposed change is to Technical Specification 3/4.9.6, " REFUELING ;

PLATFORH",- item- 4.9.6 c. , and 'is to raise the normal uptravel stops for |.the-Refueling Platform _ main and auxiliary hoists by six (6) inches, from j7 feet 9 3/4-inches to 7 feet 3 3/4 inches (measured from the fuel )

-grapple to the bottom of the platform tracks). This change vill result !in the ability-to raise fuel assemblies up to six (6) inches higher than .

currently allowed, it

2.01 Eigy.g3 for Changest

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During the Nine Mile Point Unit 2 (NMP2) 1990 Refueling Outage, vhileattempting to transfer a fuel assembly betveen the Reactor Pressure i

, Vessel and Spent Fuel Storage. Pool, a clearance problem between the '!bottom of the fuel assembly and the top of the Fuel Transfer Shield

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Bridge floor vas identified; The Fuel Transfer Shield Bridge is a- i'': temporary ' shielding device that is installed in' the ' fuel transfer slot

prior to. transferring fuel' assemblies from the reactor vessel'to-the i-

Spent Fuel Storage Pool. Attached:to the floor of-the shield bridge is iIa protective-plate that extends past the-end of-the bridge floor out-!overEthe vessel flange area,_and is intended to protect the vessel

flange area in the event of a dropped fuel assembly,,

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The clearance problem experienced was between_the bottom of the fuelassembly and the top of'the protective plate,_vhere the clearance was -

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determined to-be approximately 1/2 inch (one inch:afterLthe removal ofthe protective plate during the 1990 refueling' outage). A clearance of .;one inch does not provide a sufficient allowance for_ inaccuracy in_

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calibrating and setting the fuel grapple normal uptravel limit-stopsetpoint. With the current normal uptravel limit the possibility exists .

for. fuel assembly / shield bridge contact, resulting in possible fuel <j' bundle damage. t

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3.0 Analysis 1

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:The net effect,of the proposed change is that fuel assemblies may be _raised up to six-(6) inches higher than currently allowed during fueltransfer. The consequences-of this change ares. j

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1) -ReductLan of the minimum veter level maintained over fuel 2

assemb.ies-during fuel transfer, and i

2) Increase in the height of the fuel assembly used in the FuelAssembly Drop Analysis (USAR Section 15.7.4), i

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j 3.1 Egd ul.ign_jn Shieldi.pr Evaluation:

Evaluation of current vater level conditions during fuel transfer hasdetermined that, assuming the fuel grapple is at the normal uptravellimit (7' 9 3/4") and the Spent Fuel Pool vater level is at its minimum(352' 7 1/2"), the minimum amount of vater maintained over top of theactive fuel during the transfer is 8' 1 7/12" (8" 4 1/12" if the SpentFuel Pool is maintained at its normal vater level of 352' 10" during thetransfer). The whole body dose rate associated with the current amountof vater shielding is approximately Sur/hr, as calculated in NiagaraMohawk calculation no. 12177-PR(c)-26-K, "NMP2 Dose Rate at RefuelingPlatform due to Spent Fuel Transfer", and as measured during pastrefueling outages.

Using the same water conditions noted above, raising the normal uptravellimit six (6) inches vould provide a minimum vater shield duringtransfer of 7' ? 7/12" (7' 10 1/12" if the Spent Fuel Pool is maintainedat its normal vater level). Per Niagara Moha n calculation 12177-PR(c)-26-K, a water shield of 7' 7 7/12" vould result in a whole bodydose rate of approximately 9 mr/hr (cpproximately 6.5 mr/hr with a 7' 101/12" vater shield).

Although this change could result in a slightly increased dcse rate tothe refueling bridge personnel, the projected accumulated dose peroutage per person vould remain far belov 10CFR Part 20 1.imits. NiagaraMohawk considers the potential slight increase in dose to be acceptablein order to provide the increased clearance margin between the fuelbundles and the transfer shield bridge, and the resultant reducedpotential for possible fuel damage during transfer operations.

3.2 increaseilngL4psembly Dto.p Height EvaluaM_o_n:

Saction 15.7.4 of the NMP2 USAR provides the analysis of the mostlimiting fuel handling accident (i.e. the largest number of failed spentfuel rods) which is the drop of a spent fuel assetbly from its maximumlifted neight (approximately 30 feet) onto the reactor core. Thennalysis demonstrates that this accident vill result in radiologicalconditions far below the limits of 10 CFR 100.

The proposed change vould increase the fuel assembly drop height to amaximum of 30' 6", which represents an increase in the analyzed dropheight of less than 2%. This corresponds to a very small increase inthe kinetic energy at impact (of the dropped assembly). This smallincrease in the impact energy has been evaluated utilizing theanalytical methods described in USAR Section 15.7.4. This evaluationconcluded that the maximum possible number of fuel rods that could failfrom the impact of a dropped assembly vould be increased by 1, for atotal of 115 instead of the 124 analyzed in the USAR. The radiologicalinventory released from one additional failed fuel rod would not besignificant when added to the total released inventory from the 124failed rods analyzed in USAR Section 15.7.4, <tince the 124 failed rodsis stated in USAR Section 15.7.4 to have radiological consequences farbelow the limits of 10 CFR 100. Therefore, the proposed change villhave no significant effect on the radiological consequences of theanalyzed accident, which would still remain below the limits of 10 CFR100.

In addition, the General Electric Standard Application for Reactor Fuel(NEDE-24011-P-A-10-US, March 1991) contains a revised analysis for the

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Fuel Handling Accident. This-analysisindicatesthatamaximumofh,04.

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fuel rods could fall as'a result:of dropping a fuel assembly a distance.

of 32.95 feet. This is-less than the 124 failed rods calculated in theNMP2'USAR Chapter 15 analysis.- The improved results in-the-revised ,

-analysis are attributable to the: improved'nethodology-used in NEDE-'-

; . -24011-P, and is_a further indication that the proposed change vould haveno impact on the radiological conditions of the Fuel Handling Accident.-

- 4 . 0__ qqnclusitat. ;

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The proposed change to raise the Refueling Platform main and auxiliaryhoist normal uptravel limit _ stops by six (6) inches is needed to provide'

additional clearance between-the lifted' fuel assembly and'the top of the.

Fuel: Transfer Shield Bridge. This additional clearance is necessary to.

minimize the possibility of fuel assembly / shield bridge contact duringfuel transfers, which could result in fuel assembly and/or fuel claddingdamage. .

The radiological effects of the proposed = change have been evaluated anddetermined to be' negligible for the radiological consequences of a

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dropped spent. fuel assembly. Also, the reduced shielding over activefuel (of the liftsd assembly) has been compared to NMP2 calculation,

12177-PR(c)-26-K,_and the resulting projected slight increase _in.

operator dose has been determined to_b_e acceptable, when compared to the~

reduction in the potential for possible fuel damage during transfer-operations. Therefore,ithe NHP2 unit may be operated-safely with the-

implementation.of the proposed change.

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No Significant Hazards Analysis -

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10 CFR 50.91. requires that at-the time a-licensee requests an amendment,Lit:must provide to the Commission.its; analysis, using the: standards in-Section 50.92, about-- the issue ~ of no significant hazards consideration. .

L Therefore, in accordance with 10 CFR 50.91 and 10 CFR 50.92,'thefollowing analysis:has been performed::U

Thg_qpstafion of Nine Mile Point Unit'2. in aq,gordance~vith t,he oroposesl -

Amgndm_ent. vill not involvi s sien dicant inctggse in the probability ors;9nseauences of an acejstant previqusiv evg_lyated.

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The proposed change vill increase by.six'(6)-inches-_the maximum height~

thnt a fuel assembly can be lifted to during fuel transfers. 'The' drop#f a spent fuel assemblyLfrom its maximum.iifted height (over the. reactor core) is the most limiting fuel hcndling accident, as analyzed

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_ in' hMP2 USAR Section 15.7.4. : An evaluation, as described above, of the._ proposed chango has determined that'the additionel_6 inches vill not.

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significantly af fect the_ impact loads: analyzed in' USAR Section 15.7.4..Except for:the maximum' lifted height and' associated: limit switch

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, modification, the proposed change vill.have no other effect on thelifting mechanisms or-methods used in lifting fuel assemblies during-fuel transfers.. Therefore, neither the probability nor the consequences;of the dropped spent fuel assembly accident, as analyzed in USAR Section

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, 15.7.'4.,-vill be significantly increased.

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The operation of Nine Mile Point UniL2, in acentdance with the progandamendment. vD1not..criate the possibility of a nev QLd11Lqrent kind of

' accident frr .pnv accidsnt oreviously evaluate (x ,

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-The proposed change has been evaluated with respect to ALARAconsiderations of both the cdditional height from which a spent fuelassembly can be dropped, and the reduction in shielding (vater coveringthe active fuel in a lifted assembly). For the fuel assembly drop

accident the radiological consequences of the proposed change have been-

determined to be negligible. The projected slight increase in operatorwhole body dose rate has been determined to be acceptable, when comparedto the reduced potential for possible fuel damage during transferoperations. Except for the increase in maximum lifted height of a fuelassembly and the associated limit switch modification, the proposedchange vill not modify any fuel lifting hardware or methods used inlifting fuel assemblies during fuel transfers. The proposed change villnot alter the transfer path (between the reactor vessel and the spentfuel pool) of a lifted assembly. Therefore, the proposed change vill.

not create the possibility of a new or different kind of accident fromthe fuel handling accident analyzed in USAR Section 15.7.4.

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Ihg_.0DarA110a gf Nine lillgJ_g1Rt_Vnft 2. in accordance with the Drqpp. igdpmensiment . Vill nat_javglyA A_gigg[ficant reduction in a garrin of

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The proposed change has been evaluated and determined to have no

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significant effect on radiological cons.iderations resulting from theincrease in fuel assembly lifted height. In addition, except for'the

;- increased lifting height and associated limit switch modification, thisproposed change vill not alter the function-of the Refueling Platform,the main and auxiliary hoists, the fuel grapple, or the methods used in-'

lifting and transferring fuel assemblies. Therefore, this proposed*

change vill not cause a significant reduction in a margin of safety.

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